ML20077R710

CNS Units 1 & 2 10CFR50.59 Evaluation of FSAR Change Related Edsfi Deviation 50413,414/92-01-02 (IEEE 308-1974)

ML20077R710
Person / Time
Site:	Catawba
Issue date:	12/20/1994
From:	Dickard R DUKE POWER CO.
To:
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ML20077R694	List: ML20077R691 ML20077R710 ML20077R712
References
NUDOCS 9501230089
Download: ML20077R710 (14)

Text

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Arorm 4t.o?7$ twP7 26-941 ATTACIDENT "A" - 19 CFR 50.59 (Attachewnt 1/ s a Aiteched oocum ,ntet;,,

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1. 10 CFR:50.59?EVAliUATIONP t ,

s@efdr%ssbsfo;9p (1) STAT;ON(s): (2) UNITS (s): (3) TYPE OF ACTIVITY: O Conditionei Ope,ebiiity 0 Oconee nucloor Station Y Ua i O Nuclear Station Modification O Test or Experiment O McGuire Nuclear Station Unt2 O Minor Modification YDBD or FSAR Change Cotewba Nucteer Station O uroi 3 o e,oceoo,e o rompo,e,y Modification O O o othe, (4) DOCUMENT NUMBER / DESCRIPTION: b$k Med M NSY & On

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1. Is the actwity being evaluated a procedure, test, experiment, or evolution? If "No." proceed to Part (6). If "Yes.- O ve INo contenue to the nort question.

2. Does the stem invulvo enfrequently performed tests or evolutions that have the potential to significanth degrade, the level O ve O No of nuclear safety? If *Yes," cor sult with the Supenntendent of Operations to determine tf additional controls are necessary.

Procedure Reviewer Date:

Supenntendent of Operations Date:

mmy mn > ngwwnmn w :& , .m.m ..r.~u;;m Mpww eh JgH ~N AS,AFETYfANALYSIS' REPORT {DOCUMEWCm*:ewwyu JREVIEWN9.ga

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1. Will technical specification changes be required? O Yes No

• If the answer is "Yes,' then the chen0e may not be performed under 10 CFR 50.59.

2. TECHNICAL SPEClflCATIONS CONSULTED: 33I./dI . 3.I./.7. r3.S.2.l. .I.E.2.2.

r I.I.3.l. I.Y.3.2. r 3.I.M.

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3. FSAR SECTIONS CONSULTED: r 3./0 r

. 3.//. 8.3. r/./.2.2. f.3.2./.2. r L 3.?.M. r 15 0

4. OTHER SAR DOCUMENTS CONSULTED: SLC.j SER

5. DOCUMENT SECTIONS WHICH NEED REVISION: F56R sediens T.3./,/.z.'2. on./ g.3.2.2.4/.

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_ USQ1 EVAL.UATI,Opl.e@2 - 64 MNSDJ2093Oj33 Could the activity:

1. Increase the probabihty of an occident evaluated m the SAR? O Yes N No

2. Increase the consequences of en accident evaluated in the SAR? O ves YNo

3. Create the possibility for en occident of a dif ferent type then any evaluated in the SAR? O yes 6No 4 Increase the probability of a malfunction of equipment important to safety evaluated in the SAR? O ye ENo

5. Increase the consequences of a malfunction of equepment important to safety evaluated in the SAR? O yes INo

6. Create the possibihty for a malfunction of a different type than any evaluated in tho SAR? O Yes YNo )

7. Reduce the meroin of saf ety as defmed en the basis for any technical specification? O Yes No

' Provide en attachment to justify all answere. See NSD 209.10.4 for guidance.

• If any onewer in this section is *Yes.* the change may not be periormed under 10 CFR 50.69. .

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Anachment l' .

10 CFR 50.59 Evaluation IEEE 308-1974 FSAR Change ,

l> Page 1 of 10  ;

a PurDOSe The purpose of this evaluation is to determine the presence or absence of an ,

Unreviewed Safety Question due to the FSAR changes described below. The '

criteria of 10CFR50.59(a)2 will be applied to make this determination. This evaluation is performed in accordance with Nuclear System Directive 209 (Reference 18).  ;

Activity Descrintion During the EDSFI Inspection, February 92, a deviation was cited (Reference 1) on certain breakers in the 600 VAC Essential Auxiliary Power System (EPB) and the 125 VDC Vital Instmmentation and Control Power System (EPL). The concern was that the incoming breakers to the essential 600 VAC motor control centers are

• not coordinated with the outgoing breakers from the motor control centers and the i 125 VDC vital instrumentation and control power molded-case breakers in the  ;

distribution centers are not coordinated for all faults. i Engineering condu:ted a study of alternatives (Reference 3) which provide a resolution to this deviation. This study included a review of potential i

. modifications that could provide more coordination for these systems. The study also reviewed what types of faults would be required to cause a miscoordination ,

event and how well these systems are protected against such faults. The review of  !

potential modifications showed no breakers commercially available at the time that  !

coordinated better with downstream breakers and could be installed in existing cubicles. The conclusion of the study for both the EPE and EPL Systems was that  ;

there was not sufficient justification for major system redesigns and that FSAR  ;

changes would be pursued.

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' Attachment I >

10 CFR 50.59 Evaluation - r IEEE 308-1974 FSAR Change ,

Page 2 of 10 Proposed Revision (See attachment 2 for current wording of both sections.)

Revise FSAR section 8.3.2.2.4 as follows:

"The design of Class IE de power systems complies with the requirements of IEEE 308-1974 as augmented by Regulatory Guide 1.32. Although full coordination between protective devices is one of the primary design objectives, .

the term "to limit degradation" from section 5.2.1(6) ofIEEE 308-1974 is not interpreted to require that full coordination in all cases is the minimum requirement. Protective devices that selectively guard against some but hot all  ;

potential faults have been justified in applications where there is a very low fault probability, there is no impact to the plant safety analysis and redundant equipment is unaffected.

The Class IE batteries are given a service test at an interval not to exceed 18 l months. Additionally, the Class IE battery perfonnance and acceptance tests i comply with Section 5 ofIEEE 450-1975 and/or section 6 ofIEEE 450-1980." i Revise FSAR section 8.3.1.1.2.2 as follows: i

" Protective devices on the 600 VAC Essential Auxiliary Power System are selected and set so that a minimal amount of equipment is isolated from the system for adverse conditions such as a fault. Protective devices that selectively guard  !

against some but not all potential faults have beenjustified in applications where there is a very low fault probability, there is no impact to the plant safety analysis  !

and redundant equipment is unaffected. The load center breakers are set to protect the cable feeding the essential motor control centers and coordinate with essential motor control center feeder breakers. The relays on the essential load center transformer feeders are set to protect the transformers and coordinate with the load center breakers." .

Safety Review The 600V Essential Auxiliary Power System (EPE)(Reference 11)is provided to l supply essential power through load centers to 600V essential motor control centers. 600V essential power is supplied to nuclear safety related auxiliary

,: equipment required to maintain safe reactor status during the following plant conditions or design basis events: normal operation, hot and cold shutdown,

i

.' ' Attachment i 10 CFR $0.59 EvaluaCon IEEE 308-1974 FSAR Change Page 3 of 10

- refueling, safety injection actuation, blackout conditions, safety injection actuation ,

coupled with blackout, loss of normal feedwater, seismic event, steam generator l tube rupture, main steam line break and inadvertent opening of a steam generator ,

relief or safety valve.

The 125 VDC Vital Instrumentation and Control Power System (EPL) (Reference

10) is provided to supply power to nuclear safety related instrumentation and control loads requiring an uninterrupted power source to maintain safe reactor status during the following plant conditions: normal operation, blackout or loss of ,

offsite power (LOOP), design basis events (DBE's) including but not limited to main steam line breaks, steam generator tube rupture accidents and loss of coclant

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accidents and blackout or LOOP concurrent with one of these DBE's.

The locations for faults postulated for these systems are identified in Reference 7.

Reference 7 also includes information about Duke and industry experience with these types of faults. If one of these postulated faults results in a miscoordination event power will also be lost to some upstream loads. This would result in loss of various safety related system functions, depending on the location of the fault.

Reference 7 identifies affected systems for various postulated faults.

Any one of these postulated faults could only affect a portion of the safety train in which the fault occurs. Catawba's Fire Protection Plan (Reference 13) provides for protection of structures, systems and components important to safe shutdown such that one train of systems necessary to achieve and maintain hot standby conditions is free from fire damage. The breaker and fuse coordination taken credit for in the Fire Protection Plan is unaffected by the questions on breaker coordination covered by this 50.59 evaluation.

Unreviewed Safety Ouestion Evaluation Assumptions

1. This 50.59 evaluation is written from the perspective that these systems are being intentionally changed from a status of coordinated to the present state of" selected coordination" as defined in the-suggested FSAR changes. Since no physical changes are actually being made to the plant this perception is required to define two states of existence from which a change can be construed. This is consistent with the NSAC-125 guidance document, section 4.1.1.

i Attachment I l') CFR 50.59 Evaluation  ;

IEEE 3081974 FSAR Change i Page 4 of10 ,

2. Most of this 50.59 evaluationbddresses only Unit I circuits and equipment. The analysis is the same for the corresponding Unit 2  :

circuits and equipment.

1. Could the activity increase the probability of an accident evaluated in the SAR7 j The FSAR change reduces the level of bieaker coordination from a more protected design to the current situation with the uncoordinated fmits as identified in ,

Reference 7. The EPE system is not identified as an accident initiator should an uncoordanted fault occur. However, ifit is assumed that loss of 1EMXG is an

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accident initiator, it can be evaluated a= follows:

The PRA evaluation (Reference 9) has determined the increase in the failure '

probability of MCC IEMXG, which contains the only uncoordinated postulated fault in the Unit i EPE System. This value. 3.5E-4, is determined to be so small that when added to the expected frequency of an ANS Condition 11 Event of

- approximately 1-3 events per year, the resulting frequency is not changed to a i higher frequency category. Also, the change is small enough to not cause a

- discen. .ble increase in the probability of an accident within the same categoiy.

Analyzing the EPL System reveals that an uncoordinated fault involving loss of Distribution Center EDA, Distribution Center EDD, Panelboard EPA or i Panelboard EPD will result in a Loss of Nonnal Feedwater (FSAR Section 15.2.7) and a Reactor Trip. Thus, EPL is identified as an accident initiator. A quantified change in probability of failure of the above busses due to lack of breaker - '

coordination is not available. Reference 9 assumed a probabinty of 1.0 for the purposes of a sensitivity analysis to Core Damage Frequcacy. Therefore, an observation of historical experience for Catawba will be made. No faults resulting  !

in protective actions being required and a miscoordination event occurring have been reported at Catawba.  ;

As noted in Reference 9 (PRA evaluation) McGuire Nuclear Station has experienced one such fault. This fault occurred in September 1987 and is  ;

discussed in NRC Information Notice 88-45. This incident involved the loss of nonessential power to a portion of the main turbine control system which generated a turbine trip and subsequent reactor trip. The incident was initiated by a fault in a nonsafety compressor motor. No reports have been located 6 ?ing with a miscoordination incident on a safety related power system at either McGuire or Catawba Nuclear Station.

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Attachment 1 -

10 CFR 50.59 Evaluation ,

IEEE 3081974 FSAR Change Page 5 of10 Reference 9 reports an expecte'd frequency of such faults of 5.0E-2 per reactor year. This number (1 fault in 20 reactor years) was derived considering operation ,

of both units at both plants (Catawba and McGuire) through 1991 and 1989  ;

respectively. Considering the total experience of both plants to date (~36 reactor years) indicates an expected frequency of 2.8E-2 (1/36) such faults per reactor  ;

year. It is assumed that all faults resulting in protective action being required and  ;

a miscoordination event occurring provides an upper bound to the increase in i accident frequency due to reduced breaker coordination. Following the argument  ;

presented above for the EPE System, the expected frequency of a Reactor Trip ,

would increase from 1-3 events per year to 1.03-3.03 cvents per year. Similarly, this frequency is still in the same ANS Condition Il Category. Additionally, this -

increase is not considered a discernible increase within the same category, i This FSAR change and supporting 50.59 evaluation focuses on the issues related to a random fault resulting in unnecessary loss of equipment, instrumentation and '

control d.ie to not having " complete coordination". In some cases this postulated fault may result in the initiation of a transient (FSAR Chapter 15 Accident) such as  !

a " reactor / turbine trip" or a " loss of normal feedwater". In other cases the fault l may not be noticed with respect to plant operations but may put the unit closer to  !

trip conditions or leave the unit with less accident mitigation equipment available should an accident occur due to this fault or independently through some other means. Accident mitigation equipment will be addressed under the 50.59 l questions associated with consequences.

Referring to a letter (Reference 15) from Charles Rossi, Director of OEA/NRR, dated 5/10/89 on the NSAC-125 " Guidance Document For 10CFR50.59 Evaluations", the following conclusions can be drawn:

Reference 15 indicates that sometimes changes can be made to a plant that may )

slightly increase the probability of an accident and the changes are still acceptable with respect to the 50.59 criteria. Reference 15 also advises that a large body of ' '

knowledge has been developed in the area of event frequency and risk significant  !

sequene:s through plant specific studies and that increases in these events are  ;

I imponant insofar as they . increase the frequency of core damage. Reference 9 -

supports the position that the impact on core damage frequency has been adequately addressed and that this change in the licensing basis (FSAR change) is within the guidelines of Reference 15. That is, the expected increase in core damage frequency is low enough to complement the argument that any theoretical increase in frequency of an ANS Condition 11 accident is not discernible. i 1

' Attachment i 10 CFR 50.59 Evaluation ,

IEEE 308-1974 FSAR Change Page 6 of10

2. Could the activity increase the connequences of an accident evaluated in the SAR7- .

No, the consequences of accidents evaluated in the SAR are not impacted by this FSAR change. The accident analysis assumes one complete train of safety related equipment is not available for accident mitigation purposes. The unavailability of ,

one complete train of accident mitigation equipment, for example, can result from i the failure of a Diesel Generator during an accident sequence involving a " Loss of Offsite Power" (FSAR Section 15.2.6). The issue of breaker coordination does not  :

change these assumptions as no common mode failures (failures which can affect  !

both redundant trains) are identified. The single failure analyses presented in Table 8-8 for the "Onsite Power System" and Table 8-10 "125 VDC Vital 1 & C Power System" are not compromised by this FSAR change.

3. Could the activity create the possibility for an accident of a different type than any evaluated in the SAR7  ;

No. The only concern with lack of breaker coordination is the potential tripping of the upstream breaker rather than the breaker closest to the fault. Should this occur, the effect is "non-minimized loss of equipment". No accidents different than those previously evaluated can occur. Since faults have already been evaluated which can make unavailable as much as one channel of the EPL System (Table 8-10) and one entire train of the 4160 volt switchgear (Table 8-8) with bounding results, no i' new accidents are created.

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4. Could the activity increase the probability of a malfunction of equipment important to safety evaluated in the SAR?

Referring to the NSAC-125 guidance document for 50.59 evaluations (Reference l l

14), an increase in the probability of a malfunction of equipment important to safety evaluated in the SAR is deemed to exist if either of the following is true:

a) a change that degrades below the design basis the performance of a safety system assumed to function in the accident analysis.

b) a change that increases challenges to safety systems assumed to function in the accident analysis such that safety system performance is degraded below the design basis without compensating effects.

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"4 ,

Attachment !

10 CFR 50.59 Evaluation IEEE 308-1974 FSAR Change Page 7 of 10

' Note: A challenge to a safety system requires that safety system to perform.

Therefore, the unavailability of a safety system,if not also an accident initiator, does not challenge that safety system. For example, loss of DC power to a channel of a protection system (SSPS), may reduce the subsequent trip logic from 2/4 to 1/3, but not interrupt unit operations. This is not viewed as a challenge to safety systems. Ilowever, loss of power to the Feedwater Containment Isolation valves will cause a " Loss of Normal Feedwater" and therefore, challenge safety systems (e.g. Auxiliary Feedwater).

With respect to (a) above, this FSAR change does not degrade, below the design basis, any safety systems assumed to function in the accident analysis since there are no common cause failures which would affect both trains of any safety system.  !

Minimum safeguards assumptions are made in the accident analysis which assumes one train of each redundant safety related system is not available for accident mitigation purposes. This is a requirement of the " single failure l criterion".

With respect to (b) above, this FSAR change does not increase the challenges to safety systems assumed to function in the accident analysis such that safety system performance is degraded below the design basis. As discussed in the answer to question #1, the increase in the probability of an accident has been evaluated as not discernible. Furthermore, a fault in the EPL System that does not initiate a transient would not challenge safety systems.  !

Reference 9 quantified the increase in probability of failure of MCC IEMXG (EPE System) due to uncoordinated cor.ditions and assessed this increase as insignificant. An increase in failure probability of 3.5E-4 per reactor year was  ;

calculated and when compared to the expected failure probability of 5.1E-2 per reactor year, this represents an increase of 0.7%; which is evaluated as insignificant. It should be noted that no components powered by IEMXG are accident initiators and all components are recoverable (Reference 9). Therefore, a fault affecting IEMXG will not necessarily challenge any safety systems.

5. Could the activity increase the consequences of a malfunction of equipment important to safety evaluated in the SAR?

No. The consequences of a malfunction of equipment important to safety are not increased based on the discussion in Questions 2 above and 6 below. In order for

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• i Attachment 1 10 CFR 50.59 Evaluation IEEE 308-1974 FSAR Change Page 8 of 10 the consequences of a malfunction of equipment to be affected, a new unbounded failure mode would have to be created. As demonstrated in the responses to ,

Questions 2 and 6, this has not occurred.

6. Could the activity create the possibility for a malfunction of a different type i than any evaluated in the SAR7 No. Lack of full breaker coordination in the EPL and EPE systems could  ;

potentially result in the loss of power to a 125 VDC distribution center or a 600 VAC motor control center, which are already evaluated in the FSAR. in Tables 8-8  :

and 8-10. Therefore, no new malfunctions are created. ,

7. Could the activity reduce the margin of safety as defined in the basis for any technical specification?

No. Margin of safety defined in the bases to the Technical Specifications is ultimately related to the confidence in the fission product baniers. Without a breach of a fission product barrier, whether it be fuel pellet melting from <

exceeding centerline fuel temperature limits or cladding integrity breakdown from  :

excessive DNB, no source term would exist which could threaten the health and safety of the public. Even with fuel damage, adequate performance of the ,

remaining fission product barriers (reactor coolant pressure boundary if not a l LOCA, containment boundary and filtration systems) can greatly limit the event in terms of dose. It is from this perspective of the " defense in depth" concept that the  ;

Technical Specifications are based.

The relevant Tech. Specs. for this change are 3/4.8.1 AC Sources ,3/4.8.2 DC  !

Sources and 3/4.8.3 Onsite Power Distribution. The underlying assumptions in the j bases for these specifications are having at least one redundant set of onsite AC l and DC power sources operable during accident conditions following an assumed l

" Loss of Offsite Power" and a single failure of the other source. During periods of i time when one D/G is inoperable there is an additional action to verify that all  ;

required systems, subsystems, trains, components and devices that depend on the remaining operable D/G are also operable. This requirement provides assurance that a " Loss of Offsite Power" will not result in a complete loss of safety function j of critical systems.

A complete loss of critical safety functions can occur due to the existence of common mode failures (failures that defeat both redundant trains related to a 2

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. - Attachment I i 10 CFR 50.59 Evaluation  :

1 IEEE 308-1974 FSAR Change Page 9 of 10 particular function) or a single failure of an SSC during a time period when one of two redundant SSC's is already unavailable (beyond design basis) and the l associated action statement is governing the status of the function (since the LCO {

for that particular component can not be met! ' ingle failures are not assumed for l l

i SSC's for which an action statement is governmg further plant operations and the associated allowed outage time was derived commensurate with the importance of the particular SSC that is inoperable and the level of degradation.

This FSAR change does not affect any of the assumptions or implications in the bases to these applicable Technical Specifications. No common mode failures will result from a lack of full breaker coordination. No physical changes to the plant are being made; therefore no fission product barriers are affected. The accident analysis endpoints that describe the performance of the fission product barriers (e.g. DNB, dose consequences) is also not affected when considering the i difference in performance of the electrical protective devices (breakers not  :

minimizing loss of equipment due to miscoordination events) between having  !

" complete coordination" and having the uncoordinated faults identified in  !

Reference 7. No safety limits, setpoints or limiting safety system settings are  ;

affected by this FSAR change and no changes to any of these parameters (Tech. i Spec. changes) are required to facilitate this FSAR change without impact to any accident analyses. Therefore, the margin of safety defined in the bases to the i Technical Specifications is not reduced. 1 i

l Conclusmn l This evaluation concludes no Unreviewed Safety Questions are created by this change. FSAR changes are required and outlined on page 2. No Technical Specification changes are required.  ;

References

1. NRC Inspection Report NOS. 50-413/92-01 and 50-414/92-01 dated March 18, 1992.

2. Catawba Reply to a Notice of Violation and Notice of Deviation dated April 16,1992. u

3. Catawba Followup Report on Evaluation of Options to Resolve Deviation.

4. NRC Generic Letter 88-15 " Electric Power Systems - Inadequate Control Over i

Design Pracesses".

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Attachment i 10 CFR 50.59 Evaluation IEEE 308-1974 FSAR Change Page 10 of 10

5. April 2,1980 letter from T.A. Ledford on Duke Power Position on NRC Generic Letter 88-15.

6. IEEE 308-1974 "IEEE Standard Criteria for Class IE Power Systems for Nuclear Power Generating Stations".

7. March 2,1994 letter from D.L. Rehn to U.S. Nuclear Regulatory Commission summarizing February 7.1994 meeting between Duke Power Company and NRR.

8. September 14,1994 letter from R.E. Martin to D.L. Rehn on Request For Additional Information.

9. CNC-1535.00-00-0007, " Breaker Coordination Evaluation for the 125 VDC Vital I&C Power System (EPL) and the 600 VAC Essential Auxiliary Power

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System (EPE) (Attachment 3).

10. CNS-106.01-EPL-0001, Design Basis Document for EPL System.

I1. CNS-112.01-EPE-0001, Design Basis Document for EPE System.

12. CNS-1435.00-00-0002, Design Basis Specification For Post Fire Safe Shutdown.

13. CNS-1465-00-00-0006, Plant Design Basis Specification For Fire Protection.

14. NSAC-125," Guidelines For 10 CFR 50.59 Evaluations", June 1989.

15. May 10,1989 letter from Charles E. Rossi to Thomas E. Tipton.

16. CN-1702-05.02, EPE system one line diagram.

17. CN 1705-01.01, EPL system one line diagram.

18. Nuclear System Directive 209,"10CFR50.59 Evaluations"

19. Catawba Nuclear Station FSAR, through 1993 update.

20. Catawba Nuclear Station Technical Specifications, through amendments 125/119.

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l ATTACHMENT 2 PAGE 1 l 8.3 Onsite Power Systeu *'* * *

2) SIG (Ground Overcurrent) Coordinated with worse case 500 or SIG ,

. 3) 50B (Breaker Failure OC) 120% Bus Full bad Current

4) 62B (Breaker Failure Timer) (Breaker Operation Time) + (50B Reset Time) + 2 Cycles

11. Motor Protection ANSI Number and Function Setting

1) 50/51 (Instantaneous / Time 50-1.73 X Locked Rotor Current Delayed OC) 51 125 to 200% Full Load Current (Coordinated with Motor Start:ng and Thermal Damage Curve)

2) 500 (Ground Overcurrent) 5 Amps at 6 Cycles 111. Transformer Pmtection ANSI Number and Function Setting i) 50/51 (Inst./ Time Delayed OC) 501.73 X Maximum Low bide Fault Current 51 125 to 150% Transformer Full Load Current

2) 50G (Ground Overcurrent) ~ $ Amps at 6 Cycles To avoid protective relay trip setpoint drift problems, all Class IE relays are tested periodically to verify (

the relays are within specified limits and are re-calibrated if required. \.

L B.3.1.1.2.2 600VAC Essential Auxiliary Power System ,

2 Note:

2 Ris section of the FSAR contains information on the design bases and design criteria of this 2 system / structure. Additional information that may assist the reader in understanding the system is 3 contained in the design basis document (DBD) for this system / structure.

The 600VAC Essential Auxiliary Power System supplies power to the 600 volt essential motor control centers which are located in load concentration areas throughout the plant. Connected to the essential motor control centers are all of the 600 volt essential loads which require power during accident conditions and non essentialloads which are required to be disconnected during accident conditions. Two essential motor control centers (IEMXG and 2EMXH) are provided to supply power to loads which are shared between the two units, e.g. Control Area Chilled Water System. He Train A loads, fed from motor control center IEMXG, are identified in Table 8 6 in the remarks column. The corresponding Train B loads are fed from 2EMXH. His system is shown on Figure 8 21.

De only non Class IE loads which can be powered from the Class IE AC systems during an accident ast the AC emergency lighting transformers and the hydrogen igniter transformers. These loads are automatically disconnected on a LOCA signal and are given a permissive signal which allows manual connection after all LOCA loads are sequenced on. He AC emergency lighting transformers are powered from 600 volt Class IE motor control centers IEMXA, IEMXJ,2EMXA and 2EMXJ. The hydrogen /

igniter transformers are fed from 600 voit Class IE motor control centers IEMXI, IEMXB,2EMXI s.nd \

, 2EMXB. ,

I 8-24 (01 OCT 1993)

ATTACIIMENT 2 PAGE2 Catawba Nuclear Station FOR 10 CFR 50.59 (IEEE 308 - 1974)8.3 onsite Power Systems j k The 600VAC Essential Auxiliary Power System is divided into two redundant r.nd independent safety trains, each on which consists of two load centers and their associated motor control centers. Each load center normally receives power from its associated 4160 volt essential switchgear via a separate 1500KVA, 4160/600 volt essential load center transformer. The two load centers in each safety train are provided with a spare transformer which can be manually connected to either load center should the normal load center transfr;mer be unavailable. A key interlock scheme is provided to prevent the spare transformer from being connected to both load centers simultaneously.

In the event of a blackout or blackout coupled with a LOCA, the diesel generator load sequencer automatically sheds the load centers by tripping the load center incoming breakers. Essential loads t required during the blackout or blackout /LOCA condition are then automatically sequenced onto their i respective bus by the sequencer.

Protection devices on the 600VAC Essential Auxiliary Power System are selected and set to achieve a selective tripping 'cheme so that a muumal amount of equipment is isolated from the system for adverse conditions such as a fault. The load center breakers are set to protect the cable feeding the essential motor control centers and coordmate with essential motor control center feeder breakers. The relays on the essential load center transformer feeders are set to protect the transformers and coordinate with the load center breakers.

He protective relay settings for essential systems / equipment are calculated based on equipment manufacturer's data and system parameters. The initial setpoints are verified during system p.e operational testing. The setpoints are determined as follows:

ANSI Number and Function Setting

(

1) load Center incoming Breakers Set to protect the cable and equipment and coordinate with feeder breaker settings.

2) Imad Center Feeder Breakers Set to protect the cable and equipment.

To avoid protective relay trip setpoint drift problems, all Class IE relays are tested periodically to verify the relays are within specified hmits and are re-calibrated if required.

Refer to Section 8.3.1.4, " Independence of Redundant Systems" on page 8-39 for a description of the sepeation of redundant equipment in the 600VAC Essential Auxiliary Power System and to Section 8.3.1.3, " Physical Identification of Class IE Equipment" on page 8 38 for a detailed desenption of the physicalidentification of safety related equipment, ne instrumentation and control power for each redundant train of the 600VAC Essential Auxiliary Power System is supplied from the corresponding train of the 125VDC Vital Instrumentation and Control Power System as shown in Table 8 7. For a further discussion of the 125VDC vital system, refer m Section 8.3.2.1.2.1, "125VDC and 120VAC Vital Instrumentation and Control Power System" on page 8-44.

8.3.1.1.2.3 Testing 8.3.1.1.2.3.1 Preoperational Testing Preoperational testing of the Class IE ac system is performed in accordance with the recommendations of

, Regulatory Guide 141 to verify proper design, installation, and operation. The preoperation test program for the emergency <tiesel generators is described in Section 8.3.1.1.3.10. " Prototype Qualification Program" on page 8 33.

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ATTACIIMENT 2 PAGE 3 8.3 Onsite Power Systems *'* * " " "

FOR 10 CFR 50.59 QEEE 308 - 1974)

The 125VDC and 120VAC Vital Instrumentation and Control Power System loads are separated into I i

redundant load groups such that the loss of any one load group does not prevent the pufonnance of the '

minimum required safety functions.

He 125VDC Diesel Essential Auxiliary Power System is separated into two independent trams, one per ,

diesel. The loss of any one train does not prevent the required safety functions of the redundant diesel l generator.

Each de load group of the vital instrumentation and control systems, and each train of the diesel essential de system are supplied by a separate and mdependent battery and battery charger. The battery / battery

. charger combinations have no automatic interconnections between trains or load groups.

8.3.2.2.4 Compliance with Regulatory Guide 1.32, IEEE Standard 30819 74, and IEEE Standard 450-1975 and IEEE 450 1980 The design of Class IE de power systems complies with the requirements of IEEE 3081974 as augmented by Regulatory Guide 1.32. The Class IE batteries are given a service test at an interval nol to exceed 18 months. Additionally, the Class IE battery performance and acceptance tests comply with Section 5 of IEEE 4501975 and/or section 6 of IEEE 450-1980, 8.3.2.2.5 Class 1E Equipment Qualification Requirements The seismic and environmental qualifications of Class IE de power system equipment are discussed in Sections 3.10. " Seismic Qualification of Seismic Category 1 Instrumentation and Electrical Equipment" on page 3 214 and 3.11. " Environmental Design of Mechanical and Electrical Equipment" ' on page 3 218, respectively.

8.3.2.3 Physical identification of Class 1E Equipment The physical identification of the Class lE de systems equipment is discussed in Section 8.3.1.3, " Physical identification of Class IE Equipment" on page 8 38.

8.3.2.4 Independence of Redundant Systems ne independence of redundant Class IE de systems is discussed in Section 8.3.1.4, " Independence of Redundant Systems" on page 8 39.

TlilS IS TIIE LAST PAGE OF TIIE CllAPTER 8 TEXT PORTION.

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8-52 (01 OCT 1993)